In the late 1950's French researchers opined that tubing produced from an alloy known as Alloy 600 (nominally 72% Ni minimum, 14-17% Cr and 6-10% Fe) was susceptible to stress-corrosion attack in high purity water used in nuclear reactors. Until that time it was generally thought that the material was relatively immune to such an environment, at least in comparison with other available alloys. While there were those who considered that reactor design may have been causative of such failure, there is at least now a consensus that Alloy 600 will undergo stress-corrosion cracking with the passage of time. This in turn requires tube replacement which necessitates downtime and thus added cost.
Since circa 1960, we are aware of but one newly developed commercial alloy that has manifested an enhanced capability versus Alloy 600 to resist stress-corrosion cracking (SCC) in reactor environments, an alloy sold commercially as Alloy 690 (nominally 27-31% Cr, 7-11% Fe, 0.04% C max, balance Ni and incidental elements). Alloy 690 has gained increasing acceptance and is currently being specified as a replacement for 600 tubing. However, common to both alloys is that they are given a long time carbide precipitation heat treatment, 10-15 hours, subsequent to a mill annealing treatment. The reason for this in Alloy 600 stems from the concept of producing intergranular carbides and replenishing the area adjacent to the carbides with chromium so as to prevent sensitization caused by chromium depleted grain boundaries. As a consequence, the grain boundaries are rendered greatly less susceptible to SCC while showing no signs of sensitization.
By way of further explanation, the inner surface of tubing in respect of nuclear reactors of the high purity primary pressurized water (PWR) type is exposed to the SCC effect of the water whereas the outer surface is exposed to secondary water which may possibly contain deaerated caustic solution. The conventional 10-15 hour treatment mentioned supra provides the desired intergranular carbide precipitates thereby preventing or greatly minimizing intergranular stress-corrosion cracking of Alloy 600 in water, while cracking of Alloy 690 in water is naturally prevented by its high chromium content. This treatment also enhances both alloys' ability to resist the SCC propensity caused by the caustic solution, the effectiveness thereof being dependent upon carbon content and the mill anneal.
But long term heat treatments preclude the use of continuous annealing furnaces. Indeed as presently understood and speaking from a commercial viewpoint, there are but three current nuclear tubing manufacturers who have the necessary furnace equipment and capability to cope/deal with such long term heat treatments in the manufacture of Alloy 690 tubing. And none today is operating in the United States. Thus, the result is higher tubing costs as well as, competitively speaking, a trade disadvantage. Accordingly, the problem is one of markedly reducing the length of thermal treatment such that continuous annealing furnaces can be employed in the final sequence of operations utilized in the production of such tubing.
Given the foregoing, the problem is recognized in U.S. Pat. No. 4,336,079 anent Alloy 600. The solution described there, however, would only improve the sensitization resistance of Alloy 600 without imparting increased resistance to SCC. This is due to the formation of intragranular carbides instead of intergranular carbides. The latter are formed during the long time heat treatment and have been shown to be effective in the prevention of caustic SCC. Intragranular carbides do not afford such a benefit. It might be added that the heat treatment described in '079 would not be applicable to Alloy 690 which is not susceptible to sensitization due to its high chromium content.